Reactor Physics Calculations for the MSRE

Technical Report · Tue Jul 26 00:00:00 EDT 1960

DOI:https://doi.org/10.2172/4155392· OSTI ID:4155392

Nestor, Jr., C. W. ^[1]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

A compilation is presented of results obtained to date from a number of reactor physics calculations for the molten salt reactor experiment (MSRE). Included are one-dimensional multigroup and two-dimensional two group calculations of critical mass, flux, and power density distributions; gamma heating in the core can, reactor vessel, and core support grid; drain tank criticality; and an estimate of the beta, gamma, and delayed neutron dose rates due to fission products in the fuel contained in the pump bowl. For a cylindrical core 54 in. in diameter and 66 in. high, graphite-moderated with 8 volume % fuel salt, the calculated critical loading is 0.76 mole% uranium (93.3% U²³⁵), which is equivalent to a critical mass of 16 kg. At a reactor power of 10 mw, the peak power density in the core assuming a homogeneous mixture of fuel salt and graphite is 10 watts/cm³, the average power density is 4 watts/cm³. The computed peak thermal flux is 7.3 x 10¹³ neutrons/cm² sec and the average is 2.5 x 10¹³/ neutrons/cm² sec. Gamma heating produces a power density of 0.2 watts/cm³ in the core wall at the midplane and 0.4 watts/cm³ in the support grid at the bottom of the core at the reactor center line.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); US Atomic Energy Commission (AEC)
• NSA Number: NSA-14-023682
• OSTI ID: 4155392
• Report Number(s): CF-60-7-96
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-60
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
BETA PARTICLES
CONFIGURATION
CRITICALITY
CYLINDERS
DELAYED NEUTRONS
DIFFERENTIAL EQUATIONS
DISTRIBUTION
FISSION
FUSED SALT FUEL
GAMMA RADIATION
GRAPHITE MODERATOR
GROUP THEORY
HEATING
HOMOGENEOUS REACTORS
MASS
MECHANICAL STRUCTURES
MIXING
MSRE
NEUTRON FLUX
NEUTRONS
POWER
PRESSURE VESSELS
RADIATION DOSES
REACTIVITY
REACTOR CORE
REACTORS
THERMAL NEUTRONS
Nuclear Criticality Safety Program (NCSP)
Criticality
Molten Salt Reactor Experiment
Reactor
Salt Reactor
Gamma
Heating
Two Group Calculations
Beta Particles
Uranium
Cylinders
Cylindrical Core
Graphite
Moderation